Draft

ADVISORY BOARD ON

RADIATION AND WORKER HEALTH

National Institute for Occupational Safety and Health

REVIEW OF REVISION 03 TO THE SITE PROFILE FOR

NUCLEAR MATERIALS AND EQUIPMENT CORPORATION, 

APOLLO AND PARKS TOWNSHIP, PENNSYLVANIA

Contract No. 211-2014-58081 

SCA-TR-2017-SP001, Revision 0 

Prepared by

John Mauro, PhD, CHP 

Stephen Ostrow, PhD 

Ron Buchanan, CHP 

Joyce Lipsztein, PhD 

Joseph Zlotnicki, PhD 

SC&A, Inc.

1608 Spring Hill Road, Suite 400 

Vienna, Virginia, 22182 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

February 2017 

DISCLAIMER

This is a working document provided by the Centers for Disease Control and Prevention (CDC) technical 

support contractor, SC&A for use in discussions with the National Institute for Occupational Safety and 

Health (NIOSH) and the Advisory Board on Radiation and Worker Health (ABRWH), including its 

Working Groups or Subcommittees. Documents produced by SC&A, such as memorandum, white paper, 

draft or working documents are not final NIOSH or ABRWH products or positions, unless specifically 

marked as such. This document prepared by SC&A represents its preliminary evaluation on technical 

issues. 

NOTICE: This report has been reviewed to identify and redact any information that is protected by the 

Privacy Act 5 U.S.C. § 552a and has been cleared for distribution.
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SC&A, INC.: Technical Support for the Advisory Board on Radiation and Worker 

Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: 

Review of Revision 03 to the Site Profile for Nuclear Materials and 

Equipment Corporation, Apollo and Parks Township, Pennsylvania

DOCUMENT NUMBER/

DESCRIPTION:

SCA-TR-2017-SP001 

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: February 6, 2017

TASK MANAGER: John Mauro, PhD, CHP [signature on file]

PROJECT MANAGER: John Stiver, MS, CHP [signature on file]

DOCUMENT 

REVIEWER(S): John Stiver, MS, CHP [signature on file] 

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

0 (Draft) 2/6/2017 Initial issue
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TABLE OF CONTENTS

Abbreviations and Acronyms ......................................................................................................... 4

1.0 Introduction and Background ............................................................................................. 6

2.0 Assessment of Finding Resolutions .................................................................................... 7

2.1 Finding 1: Parks Township Site Operation Period ................................................. 7

2.2 Finding 2: Level of Uranium Enrichment for Urine Bioassay Results Expressed in 

Units of μg/L ........................................................................................................... 7

2.3 Finding 3: Guidance on Dose Reconstructions prior to 1959, and Approach for 

Missed and Unmonitored Exposures ...................................................................... 8

2.4 Finding 4: Uranium Inhalation Recommendations for the Apollo Site .................. 8

2.5 Finding 5: Inadequate Information Is Given to Replicate NIOSH’s Determination 

of Median Inhalation Concentration of Uranium.................................................. 10

2.6 Finding 6: Assumed Plutonium Mixture............................................................... 10

2.7 Finding 7: The Minimum Detectable Activities for Am-241 Lung Counting Are 

Very Low, and the Counting Method Should Be Further Explored to Give Them 

Credibility ............................................................................................................. 14

2.8 Finding 8: Air Sampling Data to Reconstruct Internal Plutonium Exposures ...... 15

2.9 Finding 9: ORAUT-OTIB-0054 Applicability ..................................................... 15

2.10 Finding 10: Internal Dose Reconstructions for RU Intakes .................................. 17

2.11 Finding 11: Helgeson Chest Count Data Concerns .............................................. 18

2.12 Finding 12: Revision of Table 6-2 and Associated Text in Section 6.3.2 of the Site 

Profile .................................................................................................................... 19

2.13 Finding 13: Better Develop Description of the Sources and Circumstances 

Responsible for External Exposures ..................................................................... 21

2.14 Finding 14: Nuclear Track Emulsion Type A (NTA) Film Adjustment Factors for 

Neutron Exposures ................................................................................................ 22

2.15 Finding 15: Non-penetrating Dose Assignments Using Film Badge Dosimeters 26

2.16 Finding 16: NUMEC Coworker Model ................................................................ 27

2.17 Finding 17: Non-penetrating Doses from Beta Emitters Associated with Residual 

Period Surface Contamination .............................................................................. 28

2.18 Finding 18: General Air Samples to Reconstruct Radionuclide Intake Rates during 

the Residual Period ............................................................................................... 29

2.19 Finding 19: Resuspension Factor during the Residual Period .............................. 29

2.20 Finding 20: Radionuclides other than Uranium during the Residual Period at 

Apollo ................................................................................................................... 30

2.21 Finding 21: Guidance on Treating Aged Plutonium Mixtures during the Residual 

Period at Parks Township ..................................................................................... 30

3.0 References ......................................................................................................................... 32
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ABBREVIATIONS AND ACRONYMS

Advisory Board Advisory Board on Radiation and Worker Health 

ALARA as low as reasonably achievable

Am americium 

AWE Atomic Weapons Employer 

B&W Babcock & Wilcox

BWXT Babcock & Wilcox Technologies 

BZ breathing zone 

Co cobalt 

Cs cesium

CWT chest wall thickness

DOE U.S. Department of Energy 

dpm disintegrations per minute

DR dose reconstruction

DWE daily weighed (average) exposure 

ER evaluation report 

FFTF Fast Flux Test Facility

FP fission product 

GA general air (sample), general area

GM geometric mean

GSD geometric standard deviation 

HASL Health and Safety Laboratory

HEU high-enriched uranium

ICRP International Commission on Radiological Protection 

keV kiloelectron volt 

LANL Los Alamos National Laboratory 

MDA minimum detectible activity

MeV mega-electronvolt 

NIOSH National Institute for Occupational Safety and Health

n:p, n/p neutron-to-photon dose ratio 

NRC U.S. Nuclear Regulatory Commission 

NTA nuclear track emulsion, Type A
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NUMEC Nuclear Materials and Equipment Corporation 

ORAUT Oak Ridge Associated Universities Team 

p/n photon-to-neutron dose ratio

PNC Japan’s Power Reactor and Nuclear Fuel Development Corporation 

Pu plutonium

RU recycled uranium

Sr strontium

SEC Special Exposure Cohort

SRDB Site Research Database

TBD technical basis document

TLD thermoluminescent dosimeter 

U uranium

UK United Kingdom

WG Work Group

ZPPR Zero Power Plutonium (later Physics) Reactor

ZPR-III Zero Power Reactor
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1.0 INTRODUCTION AND BACKGROUND 

In accordance with direction provided by the Advisory Board on Radiation and Worker Health 

(the Advisory Board) at the July 19, 2016, meeting of the Uranium Refining Atomic Weapons 

Employers (AWEs) Work Group (the Work Group, or WG) (NIOSH 2016b), this report reviews 

Revision 03 to ORAUT-TKBS-0041, Site Profile for Nuclear Materials and Equipment 

Corporation, Apollo and Parks Township, Pennsylvania, dated August 25, 2016 (NIOSH 2016c; 

also referred to as the technical basis document, or TBD), to assess if it adequately addresses the 

findings on the TBD that SC&A had developed since its initial publication in 2008 (NIOSH 

2008). The work site description section of the National Institute for Occupational Safety and 

Health (NIOSH) website for the Nuclear Materials and Equipment Corporation (NUMEC) 

provides a detailed compendium of the entire review and publication process

(http://www.cdc.gov/niosh/ocas/numec.html). 

Some of the most pertinent recent events and reports related to the findings are as follows: 

• NIOSH Responses to SC&A Review of ORAUT-TKBS-0041, Rev. 02: Site Profile for 

Nuclear Materials and Equipment Corporation, Apollo and Parks Township, 

Pennsylvania, Revision 0-A, National Institute for Occupational Safety and Health, 

Cincinnati, Ohio, May 14, 2015 (NIOSH 2015b) 

• SC&A Assessment of Selected NIOSH Responses to SC&A NUMEC Site Profile Findings, 

SC&A, Inc., Vienna, VA, August 14, 2015 (SC&A 2015) 

• White Paper on NUMEC Issues and Feasibility of Coworker Model, Revision 0, National 

Institute for Occupational Safety and Health, Cincinnati, Ohio, June 23, 2016 

(NIOSH 2016a) 

• July 19, 2016, WG meeting. 

All findings were resolved “in principle” during the July 19, 2016, WG meeting, pending review 

of Revision 03 to the TBD, which was issued August 25, 2016, and some responses by NIOSH. 

The objective of this report is to assess whether Revision 03 to the TBD does, in fact, contain the 

information and revisions agreed upon during the WG meeting, thereby allowing the WG to 

close the findings at a future meeting. The format used to present SC&A’s review of the TBD is 

to briefly summarize each of the 21 findings, outline how they were resolved in principle during 

deliberations by the WG, identify any unresolved issues, and assess the degree to which the 

findings were adequately addressed in Revision 03 to the TBD or otherwise clarified by NIOSH. 

Each section of this report (1) uses the NIOSH white paper of June 23, 2016 (NIOSH 2016a), as 

its starting point for convenience, as that paper systematically lists each finding and its status;

(2) adds material on NIOSH’s position for that finding; (3) adds any comment SC&A may have 

expressed about that finding; and (4) evaluates whether that finding has been adequately 

addressed in Revision 03 to the TBD or otherwise addressed by NIOSH. The material in italics is 

excerpted directly from NIOSH 2016a. As indicated in SC&A’s August 2015 report (SC&A 

2015), at that time, of the original 21 SC&A findings developed during the NUMEC site profile 
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review process, 11 were closed and 10 remained in progress. As described in more detail in the 

sections that follow, this current report concludes the following: 

SC&A recommends closing all 21 findings except Findings 6, 10, and 12. 

2.0 ASSESSMENT OF FINDING RESOLUTIONS 

The following subsections present the status of each of the 21 findings in turn. The italicized 

paragraphs are quoted from NIOSH 2016a, followed by SC&A’s evaluation and 

recommendation to the WG for disposition of the finding. Note that, for convenience in referring 

to the findings, SC&A has given each finding a short, descriptive subsection title. 

2.1 FINDING 1: PARKS TOWNSHIP SITE OPERATION PERIOD

Finding 1: Clarification is needed regarding the start and end dates of Parks 

Township Site operations. 

Proposed NIOSH resolution: Tables will be updated with new information 

and values possibly corrected. 

August 2015 WG discussion: Closed, if change is included in TBD revision. 

NIOSH update June 2016: Change included in draft ORAUT-TKBS-0041 

Rev. 03-B. 

SC&A Evaluation: SC&A first raised this issue in SCA-TR-SP2013-0038, Review of the 

NIOSH Site Profile for Nuclear Materials and Equipment Corporation Park Township and 

Apollo Sites, April 19, 2013 (SC&A 2013), which states that Section 2.2.1 of the site profile (at 

the time) indicates that the site was not authorized to begin work until 1961, but Table 2-4 of the 

site profile indicates that many of the locations were operational in 1960. Additionally, 

Section 2.2.3.3 indicates that Building C was not used from construction until 1973, and 

Section 2.2.5 states that all operations in Building C ceased in 1978, yet Table 2-4 indicates that 

operations in Building C began in 1969 and lasted till 1980. This issue was closed during 

subsequent deliberations but needed to be confirmed once Revision 03 to the TBD was issued. 

Inspection of Table 2-3 in Revision 03 to the TBD indicates that it has been revised to indicate 

that no facilities were operational before 1961.

SC&A Recommendation: SC&A recommends closing this issue. 

2.2 FINDING 2: LEVEL OF URANIUM ENRICHMENT FOR URINE BIOASSAY 

RESULTS EXPRESSED IN UNITS OF μg/L

Finding 2: The site profile should provide guidance about what level of uranium 

enrichment should be assumed for those urine bioassay results that are expressed 

in units of μg/L.
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Proposed NIOSH resolution: NIOSH will add a section to TBD with guidance 

on enrichment assumption if no other information is available (assume highly 

enriched since it is claimant favorable and was used at Apollo). 

August 2015 WG discussion: Closed, if guidance is included in TBD revision.

NIOSH update June 2016: Change included in draft ORAUT-TKBS-0041 Rev. 

03-B.

SC&A Evaluation: SC&A first raised this issue in SC&A 2013 regarding TBD Section 4.4.1, 

which states that for those urine analysis results expressed in terms of μg/L, no guidance is 

provided regarding what level of enrichment should be assumed. NIOSH claimed to have 

resolved this issue with a revision to the TBD that states that the dose reconstructor should use 

the claimant-favorable assumption that the uranium is enriched if no other information is 

available. 

Inspection of Section 5.2.2.4 of Revision 03 to the TBD does, in fact, state: “If the enrichment is 

not known, then an estimate of activity that is favorable to the claimant can be made by assuming 

the material to be HEU [high-enriched uranium] (93%).”

SC&A Recommendation: SC&A recommends closing this issue. 

2.3 FINDING 3: GUIDANCE ON DOSE RECONSTRUCTIONS PRIOR TO 1959, AND 

APPROACH FOR MISSED AND UNMONITORED EXPOSURES

Finding 3: Some guidance is needed on how to perform dose reconstructions 

prior to 1959, and what approach to use for missed and unmonitored exposures.

Proposed NIOSH resolution: None. Pre-1959 dose reconstruction is 

infeasible per SEC [Special Exposure Cohort]. 

August 2015 WG discussion: Closed. 

SC&A Evaluation: SC&A concurs with NIOSH’s proposed resolution; i.e., if there is no 

technical approach that can be used to reconstruct doses to workers who are not covered by the 

SEC, no dose reconstruction can or should be performed.

SC&A Recommendation: SC&A’s review of Revision 03 to the TBD does not alter the basis of 

the August 2015 WG decision to close the finding. 

2.4 FINDING 4: URANIUM INHALATION RECOMMENDATIONS FOR THE 

APOLLO SITE 

Finding 4: Uranium inhalation recommendations for the Apollo Site need to 

consider the method discussed by Davis and Strom (2008) for dealing with 

uncertainties in DWE [daily weighted (average) exposure]. This technique was 

evaluated and found to be appropriate in the site profile review for the Fernald 

Feed Materials Production Center (Fernald). 
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Proposed NIOSH resolution: Davis and Strom (2008) evaluated several HASL

[Health and Safety Laboratory] reports and concluded that the daily weighted 

exposures have a lognormal distribution with a geometric standard deviation 

of 5. HASL data is available for the Apollo site and has been used to evaluate 

exposure concentrations. As discussed in Finding 18, the air concentration 

used for the beginning of the residual period has been re-evaluated and is 

now based on General area (GA) monitoring rather than daily weighted 

average exposures (see response to Findings 5 and 18 below). The geometric 

standard deviation (GSD) for the revised air concentrations has been set to 

5.0 based on Davis and Strom (2008). 

August 2015 WG discussion: The discussion during the work group meeting 

moved from the question of using the HASL data in dose reconstructions to 

the more general issue of whether it was possible to develop co-worker 

models for uranium intakes. Based on the Apollo Evaluation Report, internal 

doses from uranium bioassay can be reconstructed when data are available 

for an individual claim. SC&A questioned if NIOSH has any claims with 

uranium bioassay where internal uranium dose was not assigned. SC&A 

would like to investigate more to find workers who were exposed to U

[uranium] and not had internal U assigned. NIOSH was tasked to review and 

evaluate if a coworker model is needed. Issue 4 was closed; however, NIOSH 

committed to assess the possibility of a coworker model. Dr. Neton pointed 

out that the SEC decision does not mean that every single feasibility and 

infeasibility was evaluated (meaning, just because the finding for the Apollo 

ER [evaluation report] was that internal uranium dose reconstruction are 

feasible when data exist for a given claim does not necessarily mean that a 

coworker model can be developed for this data). Any coworker model that is 

developed would have to hold up to the standards of the coworker 

implementation guide. 

NIOSH update June 2016: During the initial discussion, there was some 

misunderstanding in NIOSH’s part on the focus of this issue. The original 

issue addresses Section 5.4 of the TBD, which provides guidance to assign 

internal doses based on fixed BZ [breathing zone] air data for uranium 

workers during 1959-1961. Upon further consideration of the guidance and 

the data, this section using HASL BZ data will now be removed from the TBD 

in accordance with the findings presented in the Apollo SEC evaluation report 

that internal doses cannot be reconstructed during that time period except 

when uranium bioassay are available. HASL BZ data were collected for 

specific projects and operators during a limited operating time and do not 

reflect the range of uranium activities carried out and NUMEC. 

As for the coworker model discussion, it is not expected that the available 

data can be used to develop coworker models that would adhere to the 

guidelines expected from such a model. Please see detailed analysis below.
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SC&A Evaluation: This issue has a complex history, as described in NIOSH 2016a. The 

important conclusion made by the NIOSH paper is that it is not expected that the available data 

can be used to develop coworker models that would adhere to the guidelines expected from such 

a model. This revised strategy for dose reconstruction makes this issue moot. 

Therefore, SC&A concurs with this position, and the matter is considered adequately addressed 

in Revision 03 to the TBD.

SC&A Recommendation: SC&A recommends closing this issue, 

2.5 FINDING 5: INADEQUATE INFORMATION IS GIVEN TO REPLICATE 

NIOSH’S DETERMINATION OF MEDIAN INHALATION CONCENTRATION 

OF URANIUM

Finding 5: Inadequate Information is given to replicate NIOSH’s determination 

of median inhalation concentration of uranium. The NIOSH result could not be 

replicated, and it appears that relevant information has been omitted from the 

HASL studies reported in Appendix A to the site profile. 

Proposed NISOH resolution: No data from the HASL reports cited have been 

omitted, but the calculation approach has been revised since the previous 

TBD was issued and is now based on GA data. 

August 2015 WG discussion: NIOSH should provide more guidance how the 

revised approach is to be used for partial dose reconstructions. This analysis 

was done to get to a starting point for levels to be used in the residual period. 

For the operational period, BZ samples are used if they are available. SC&A 

voices concerns that there were significantly higher intakes in the ceramics 

plant than other areas and whether those would be represented in the median 

intakes. 

NIOSH update June 2016: See response to Finding 4. 

SC&A Evaluation: The publication record and Section 7.4 of Revision 03 to the TBD state that 

Table 7-3 has been revised to use general air sampling data rather than breathing zone data as the 

basis for intakes of residual uranium. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.6 FINDING 6: ASSUMED PLUTONIUM MIXTURE

Finding 6: The site profile would benefit from a discussion demonstrating that 

the Hanford fuel-grade mix, as opposed to the weapons-grade or commercialgrade plutonium, is limiting for the full range of plutonium mixes and ages that 

were used at NUMEC. In addition, given the complexity of this subject, a review 

of actual dose reconstructions would provide greater insight into how this matter 

is actually being addressed. 
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Proposed NISOH resolution: Table will be updated with recently captured 

information on the fuel types used at NUMEC to represent more accurately 

the mixes that were handled. 

August 2015 WG discussion: SC&A wants to review response in detail and 

provide comment on the response. SC&A was tasked with providing an 

additional response. 

SC&A Response after WG discussion: Other possible Pu mixes that may have 

been present at NUMEC should be investigated (such as Japanese fuels and 

fuels from the UK [United Kingdom]). 

NIOSH update June 2016: This issue was not pursued further. Additional data 

capture is not considered to be a productive option for NUMEC. The 

currently available data can be used for a reasonable claimant-favorable 

assumption to assign partial internal dose. It would not be reasonable to 

speculate on additional Pu mixtures to complete partial dose reconstructions 

for non-presumptive cancers. Any reported Pu bioassay will be used at face 

value. No adjustments for other contaminants are available. 

SC&A Evaluation: As indicated in Section 1.1 of SC&A 2015, NIOSH provided extensive 

additional information about the plutonium-grade mix handled at NUMEC by compiling 

applicable information from the Hanford TBD, and committed to include this new information in 

the next revision of the TBD. SC&A agreed that this new information essentially resolved this 

issue. However, as stated in the SC&A report, several additional questions needed to be 

addressed: 

• NIOSH’s May 14, 2015, report (NIOSH 2015b) does not appear to address all the 

possible sources and mixtures of plutonium that might have been present at NUMEC 

over the years. While it presents new information about Japanese and Zero Power Physics 

Reactor (ZPR-III, located at Argonne National Laboratory-West on the Idaho National 

Laboratory site) fuels produced in the 1960s, it does not discuss possible United 

Kingdom fuel or other fuels. SC&A requested that NIOSH address these specific 

categories of fuel, as was done for the other categories of fuel. 

• The report (SC&A 2015) asked if NIOSH will run all combinations of grades and ages of 

plutonium that might apply and use the limiting scenario to envelope the actual dose. If 

so, will NIOSH develop a workbook that can be used under these circumstances? 

• Additionally, SC&A suggested that NIOSH include Hanford Table 5-5, “Activity 

Composition of Hanford Reference Commercial Power Fuel-Grade Plutonium Mixture” 

(in addition to Tables 5-3 and 5-4), in the site profile for convenience to the dose 

reconstructor.

NIOSH’s response was reiterated at the July 19, 2016, Uranium Refining Atomic Weapons 

Employers WG meeting (NIOSH 2016b). NIOSH stated that they do not have any more data and 

that what they “currently have can be used to make a reasonably claimant-favorable assumption 

for partial dose reconstructions and anything else would be more or less speculation...We do not 
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expect that any additional data capture is possible at this site.” NIOSH also repeated its previous 

commitment to revise the appropriate section in Revision 03 to the TBD with the addition of Pu 

mix information for three other Pu-240 weight percentages (as presented in NIOSH 2015b), and 

to supply more guidance to the dose reconstructor; e.g., to ensure that the most claimantfavorable mixture is chosen by running a few different cases. SC&A accepted NIOSH’s strategy 

in general but reserved making a recommendation to the Work Group before seeing Revision 03 

to the TBD.

Inspection of Revisions 02 and 03 to the site profile reveals the following with respect to the 

degree to which issues related to Finding 6 are now fully addressed. Revision 02 (NIOSH 2012a) 

notes that there are three types of plutonium mixtures, reactor grade (27 weight percent Pu-240), 

weapons grade (6 weight percent), and fuel grade (12 weight percent), with Table 5-3 containing 

specific activities by isotope and different fuel aging times for only the fuel-grade mixture 

(“Hanford reference-grade plutonium,” which was supplied to the Fast Flux Test Facility 

[FFTF]). Table 5-3 of Revision 03. However, Revision 02 presents specific activities for those 

three mixes as well as for 8.5 weight percent, U.S. Department of Energy (DOE) plutonium, for 

a total of four mixes ranging from 6 to 27 weight percent Pu-240 in the plutonium matrix. 

Revision 03 contains an expanded discussion of the different types of plutonium-based fuels 

encountered at NUMEC as well as guidance for the dose reconstructor, which is reproduced here 

for convenience: 

Fuel-grade plutonium was used for FFTF fuel fabrication and most of the ZPPR 

fuel fabrication (11,500 plates). However, commercial reactor-grade plutonium 

was also used for some ZPPR [Zero Power Plutonium (later Physics) Reactor]

fuel fabrication (700 fuel plates) starting in April of 1969. For work in the FFTF 

fuel fabrication, lacking specific information on the actual composition of the 

processed plutonium, the composition of Hanford fuel-grade plutonium can be 

used because this was the source of plutonium for FFTF fuel fabrication (Author 

unknown 2004). The site also used Hanford 6% weapons-grade plutonium for 

some mixed-oxide fuel fabrication. An 8.5% plutonium mix was used for a 

contract with Japan’s Power Reactor and Nuclear Fuel Development 

Corporation (PNC) and for ZPR-III fuel. The activity compositions for the four 

grades of plutonium are listed in Table 5-3 for fuel aged up to 20 years ([NIOSH]

2010b, 2015[d]). The age of plutonium to assume for a given analysis depends on 

the radionuclide measured in the bioassay analysis. When gross alpha, 238Pu, or 

239Pu is measured, the dose is maximized by assuming longer decay times 

(20 years). When 241Am is measured and the intake is estimated using ingrowth of 

241Am from decay of 241Pu, the dose is maximized by assuming a short (5-year) 

decay time. If the actual age of the fuel is known (such as from an incident 

investigation report), that age can be used in the intake and dose analysis. If the 

exposure occurred only in FFTF fuel fabrication, then the Hanford fuel-grade 

composition can be used. If the work location is not known (such as for plutonium 

scrap recovery), then use of the Hanford 6% weapons-grade plutonium will 

provide a dose estimate that is favorable to the claimant. 
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While reviewing NIOSH’s response to this finding, SC&A also became aware that a recent 

NIOSH report, DCAS-RPT-005, Revision 0, Alternative Dissolution Models for Insoluble 

Pu-238, (NIOSH 2016d). NIOSH 2016d lists NUMEC as one of the sites where workers could 

be exposed to special Pu compounds (i.e., unencapsulated, insoluble Pu-238), and that the lung 

model parameters derived for Los Alamos National Laboratory (LANL) should be used. The 

document explains that the special dissolution type for Pu-238 (Super-S, or SS) should be used in 

addition to Types M and S, and that the dissolution type that results in the highest dose to the 

organ of interest should be used for the final dose assignment. SC&A confirmed that Revision 03 

to the TBD addresses this issue in Section 5.1, as follows: 

The dose reconstructor should use the solubility type that results in the highest 

dose. 

... 

In general, plutonium oxides, carbides, and hydroxides are absorption type S; 

nitrates and other compounds are type M (ICRP 1995, p. 328–329). Older 

materials, even when starting out as soluble, can have a tendency to oxidize when 

left in contact with air. Oxides, metals, and old contamination should be treated 

as type S. If nothing is known about the chemical form of plutonium, either type M 

or S can be used to maximize the dose to the organ of concern. In addition, 

because highly insoluble forms of plutonium might have been present, guidance in 

ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly Retained in the 

Lung ([NIOSH 2010c]), should be followed for the evaluation of highly insoluble 

(type SS) plutonium. Americium-241 is a component of plutonium contamination 

and should be modeled in the lung the same as the plutonium matrix in which it 

has grown. In other words, the americium should be treated as absorption type S 

if the plutonium is type S (ORAUT 2015). If the plutonium is type SS, follow 

guidance in [NIOSH 2010c] for assignment of the 241Am solubility type.

Notwithstanding the above, SC&A suggests that for completeness and to provide a full 

discussion, NIOSH should supplement Section 5.1 of the TBD by noting that ORAUTOTIB-0049 (NIOSH 2010c) is applied only to doses resulting from intakes of plutonium 

for which the activity isotopic ratio of Pu-(239+240) to Pu-238 is greater than one. When 

this condition is met, Type SS behavior applies to all isotopes in the plutonium mixture. 

In addition, when workers are exposed to basically pure, insoluble Pu-238, special 

dissolution models should be used in dose reconstruction, as specified in DCAS-RPT005, Revision 0 (NIOSH 2016d). According to this report, NUMEC workers could have 

been exposed to significant quantities of unencapsulated, insoluble Pu-238. The 

alternative dissolution model for Pu-238, with parameters defined in NIOSH 2016d, 

should be tested against the standard Pu-238 types to define which lung model results in 

the highest dose to the organ of interest. NIOSH should add an additional special 

dissolution type to TBD Section 5.1 for sources of almost pure insoluble Pu-238, as 

defined in DCAS-RPT-005, Revision 0. 

SC&A is satisfied that Revision 03 to the TBD adequately responds to the original concerns of 

Finding 6, but not to the additional issue of the solubility type. 
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SC&A Recommendation: SC&A recommends that this issue remain open pending further 

discussion of the solubility class for almost pure insoluble Pu-238. 

2.7 FINDING 7: THE MINIMUM DETECTABLE ACTIVITIES FOR Am-241 LUNG 

COUNTING ARE VERY LOW, AND THE COUNTING METHOD SHOULD BE 

FURTHER EXPLORED TO GIVE THEM CREDIBILITY

Finding 7: The MDAs [minimum detectable activities] for Am-241 lung counting 

are very low, and the counting method should be further explored in order to give 

them credibility.

Proposed NIOSH resolution: NIOSH will add some additional guidance to the 

draft TBD. 

August 2015 WG discussion: NIOSH has addressed some of the MDA issues 

and added some more reasonable values, but SC&A’s concerns remain. 

Dr. Neton agrees in that the MDA numbers for Pu look somewhat low, the Am 

MDA looks ok. The Pu MDA values are very much dependent on chest wall 

thickness. SC&A was tasked with providing additional assessment. 

SC&A response after WG discussion: SC&A believes that the MDA value for 

in-vivo monitoring of Am-241 and Pu-239 are not reliable, that very limited 

data is available, and that the low reported values for MDAs for Am-241 

in-vivo lung monitoring need to be further developed. The values for Pu-239 

are not credible, due to the 17 keV X-rays being measured directly. 

NIOSH update June 2016: NIOSH has discussed and evaluated the chest 

counting data that was supplied by Babcock and Wilcox. Many of the reported 

results have a reported MDA when a non-detectable amount is reported. The 

MDAs appear to be lower than what might be reported today, but it would be 

difficult to come up with an alternate value. Most measurements were done at 

the University of Pittsburg and the NUMEC in-vivo program was overseen by 

a person who was highly regarded in the field. The lower MDAs could be the 

result of the calibration phantoms used at the time. Until the advent of the 

Livermore realistic phantom in the late 1970’s, the phantoms in use did not 

provide a sophisticated representation of human anatomy and tissue 

equivalency. It was eventually discovered that the earlier phantoms tended to 

overestimate counting efficiency, especially for low-energy emitters like Pu 

and Am. Unfortunately, there does not seem a straightforward way to adjust 

the earlier values retroactively, to comport with what might be reported 

today. 

There are too many unknown variables involved, including the exact design 

and configuration of the detectors used, the position of the detectors over the 

lungs (i.e., measurement geometry), and the chest wall thickness (CWT) for 

each subject counted. Even if the CWT is known for a person, the manner in 

which it was measured or calculated can lead to large differences in the 
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value. Given all this, NIOSH intends to use the detection limits as reported for 

each original measurement. While there may be some negative bias 

associated with using these values, it seems that this is preferable to making 

adjustments that would largely be based on conjecture. This approach would 

be consistent with performing partial dose reconstructions for nonpresumptive cancers using “data that may be included in an individual’s file 

(and that can be interpreted using existing NIOSH dose reconstruction 

processes or procedures)” to support a partial internal dose reconstruction 

for non-presumptive cancers and/or cases that have less than 250 workingdays of employment. 

SC&A Evaluation: SC&A believes that NIOSH has thoroughly investigated this issue and has 

adopted a reasonable strategy for implementing this strategy, as indicated in the above excerpt 

from Revision 03 to the TBD. It is also noteworthy that the protocol proposed by NIOSH is a 

means for assigning at least some dose to workers who would otherwise not be assigned any 

dose under the SEC. 

Hence, SC&A believes that NIOSH has made every reasonable effort to assign a reasonable dose 

to uncovered workers. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.8 FINDING 8: AIR SAMPLING DATA TO RECONSTRUCT INTERNAL 

PLUTONIUM EXPOSURES 

Finding 8: The site profile would benefit from a more thorough discussion of the 

possible use of air sampling data to reconstruct internal plutonium exposures and 

to take into consideration the additional data provided by Crosby 1967 and 

NUMEC 1967. 

Proposed NIOSH resolution: None. Internal dose reconstructions are partial 

and done only in cases where bioassay is available. 

August 2015 WG discussion: Closed. 

SC&A Evaluation: SC&A concurs with NIOSH’s position on this matter because air sampling 

data are not used for dose reconstruction and, therefore, this issue becomes moot. 

SC&A Recommendation: SC&A’s review of Revision 03 to the TBD does not alter the basis of 

the August 2015 WG decision to close the finding. Revision 03 to the TBD does not rely on air 

sampling data to reconstruct internal dose, and, therefore, the TBD need not address this issue. 

2.9 FINDING 9: ORAUT-OTIB-0054 APPLICABILITY

Finding 9: It does not seem appropriate to use ORAUT-OTIB-0054 ([NIOSH]

2007a) to reconstruct the internal exposure of workers at NUMEC who might 

have been exposed to mixed fission products. ORAUT-OTIB-0054 states that its 

guidance “does not apply to determination of intakes where radionuclides have 
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been purposely extracted and concentrated as for heat generation sources, 

medical uses, or waste handling operations that caused significant alteration to 

the source term to which workers were exposed.” For example, see Table 5-1 of 

the site profile. Also, the fission product mix given in ORAUT-OTIB-0054 does 

not contain the same radionuclides as the fission product mixes given for the 

NUMEC Laundry in the 1975 effluent release report (SRDB #20081) and for invivo count results in SRDB #19970. The NUMEC mixes include Co-60, which the 

ORAUT-OTIB-0054 mix omits. 

Proposed NIOSH resolution: The use of OTIB-0054 ([NIOSH 2015a]) is 

appropriate when the source of activity is from reactors. As indicated in this 

finding, the NUMEC laundry processed materials from many sources in 

addition to the commercial reactor laundry. The site profile will be modified 

to remove guidance on use of OTIB-0054 for evaluation of dose from 

associated fission products and activation products. The guidance will 

indicate that doses are to be evaluated only for the radionuclides included in 

the bioassay results 

August 2015 WG discussion: Closed, if new guidance is included in TBD 

revision. 

NIOSH update June 2016: Change included in draft ORAUT-TKBS-0041 Rev. 

03-B.

SC&A Evaluation: At the Work Group meeting of August 3, 2015, the Advisory Board 

members requested that SC&A evaluate some of NIOSH’s responses to SC&A’s findings in 

NIOSH’s May 14, 2015, report (NIOSH 2015b), but did not include Finding 9, noting that the 

finding should be closed if NIOSH provided appropriate revised guidance in Revision 03 to the 

TBD. Hence, SC&A’s subsequent report of August 14, 2015 (SC&A 2015), did not contain an 

assessment of NIOSH’s response to Finding 9. 

NIOSH responded to various NUMEC issues raised by SC&A in a June 23, 2016, white paper 

(NIOSH 2016a). Its section on Finding 9 reiterates the statement contained in NIOSH 2015b and 

notes that the change was made in the draft TBD Revision 03. 

Discussions at the July 19, 2016, WG meeting covered several of the findings, but not Finding 9. 

SC&A reviewed the portion (Section 5.2.4, “Mixed Fission Products”) of Revision 03 to the 

TBD applicable to Finding 9 to assess how it addresses the issues that were raised. As promised 

by NIOSH, all references to OTIB-0054 that were present in Revision 02 of the TBD were 

removed in Revision 03. After noting radionuclides that could have been encountered at either 

the Apollo site (where the Laundry Building was located) or the Parks Township site, the revised 

TBD states:

Urine bioassay data for mixed FPs [fission products] should be used, if included 

in the case files, to estimate intakes of FPs. The mixed FP urine bioassay results 

do not indicate if the measurements are based on beta or gamma analysis, so the 
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intake should be based on 90Sr as a representative beta emitter and 137Cs as a 

representative gamma emitter and the analysis providing the highest dose should 

be used. 

SC&A is satisfied that Revision 03 to the TBD adequately responds to the concerns of Finding 9 

and recommends that the WG close the finding. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.10 FINDING 10: INTERNAL DOSE RECONSTRUCTIONS FOR RU INTAKES 

Finding 10: Internal dose reconstructions performed for NUMEC personnel 

might need to be revisited in light of changes to the Fernald site profile ([NIOSH]

2004) with respect to RU [recycled uranium]. Also, additional direction is needed 

with respect to which workers or operations should be assigned RU intakes. 

Proposed NIOSH resolution: Some language will be added to the revised 

NUMEC TBD regarding assigning RU doses for NUMEC. 

August 2015 WG discussion: Discussed whether any additional sources of 

information regarding the uranium composition were reviewed. Closed if 

NIOSH adds guidance on assigning RU doses to the TBD revision. 

NIOSH update June 2016: Change included in draft ORAUT-TKBS-0041 Rev. 

03-B.

SC&A Evaluation: As indicated in the Publication Record of Revision 03 to the TBD, revised 

Section 5.2.5 indicates recycled uranium contaminants should be included in the evaluation of 

the material solubility type that gives the highest internal dose. A review of that section of the 

TBD reveals that the NIOSH states that: “The intake of RU contaminant radionuclides can be 

estimated using the guidance in Section 5.3.2 and activity fractions in Tables 5-33 and 5-34 of 

ORAUT-TKBS-0017-5, Technical Basis Document for the Feed Materials Production Center – 

Occupational Internal Dose ([NIOSH 2016e; i.e., Revision 01]).” 

The Fernald site profile was recently revised (NIOSH 2016f; i.e., Revision 02), which, in part, 

corrected the RU values that were recommended in Table 5-34 of Revision 01 to the Fernald site 

profile. Section 5.3.2.1 of Revision 02 to the Fernald site profile acknowledges that 

“concentration of contaminants in RU changed with time, so default values for dose 

reconstruction are defined for discrete time intervals.” The recommended RU contaminant mass 

concentrations for use in dose reconstruction at Fernald are provided in Table 5-10. Two sets of 

values are provided; one set for 1961–1972 and a second set for 1973 to the present. 

Accordingly, Revision 03 to the NUMEC site profile should: (1) be revised to refer to RU values 

in Revision 02 to the Fernald site profile (as opposed to Revision 01), and (2) also provide 

guidance with respect to which RU time periods in the Fernald site profile should be used. 

Finally, a review of Revision 02 to the Fernald site profile, with respect to RU, reveals that many 

aspects of RU at Fernald are quite complex because of RU-containing plutonium that was out of 

specification. Accordingly, Section 5.2.5 of Revision 03 to the NUMEC site profile also needs to 
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address issues related to the unique aspects of RU at Fernald and the degree to which they apply 

to NUMEC.

SC&A Recommendation: SC&A recommends that this issue remain open until these concerns 

are resolved.

2.11 FINDING 11: HELGESON CHEST COUNT DATA CONCERNS

Finding 11: NIOSH should explain whether the concerns expressed in the Pantex 

site profile ([NIOSH 2007b]) about the Helgeson chest count data might also 

apply to chest count data at NUMEC performed by Helgeson for NUMEC 

workers. 

Proposed NIOSH resolution: The in vivo counting at NUMEC was performed 

by Helgeson only during 1968, 1971, and once in 1975. The Helgeson counts 

included primarily plutonium-239, with a few results for uranium and fission 

products. The majority of in-vivo counting was performed at the University of 

Pittsburg hospital. The issue with Pantex was that the counts for uranium 

were biased high and represented false positives (Brake 1989). The current 

Pantex internal dose TBD has eliminated all reference to Helgeson in-vivo 

measurements. Using the Helgeson results for uranium at NUMEC is 

favorable to the claimant, because any bias would be high. There was no issue 

identified related to Helgeson in-vivo counting for fission products, 

plutonium, or americium. 

August 2015 WG discussion: SC&A wants to look into this issue more. SC&A 

will review the issue and provide findings to NIOSH. 

SC&A response after WG discussion: SC&A remains concerned about the 

values used for MDA for in-vivo counts, as some are listed as 63 μg and some 

are listed as 80 μg for U-235. SC&A would like to see clarification and DR 

[dose reconstruction] examples to ensure the correct value is used. 

NIOSH update June 2016: NUMEC used the Helgeson mobile whole body 

counter for Pu and Am counts, and also some U-235. The MDA reported for 

U-235 in 1968 using the Helgeson unit is 80 μg. NUMEC mostly did whole

body counts at the Low Level Radiation Monitoring Facility at the University 

of Pittsburg, which, on many count results, report a MDA value of 63 μg for 

U-235. A clarifying statement could be added to the TBD, but since generally 

the MDAs seem to be reported with the respective result, for a nonpresumptive cancer, the MDA respective to the analysis would be used. As 

outlined for item 7, NIOSH has not located any information that would allow 

for a critical assessment and correction of reported MDA values that are 

perceived low by SC&A. The issue with Pantex was that the counts for 

uranium were biased high and represented false positives (Brake 1989). The 

current Pantex internal dose TBD has eliminated all reference to Helgeson in 

vivo measurements. Using the Helgeson results for uranium is not to the 
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detriment to the claimant, since it would lead to a positive bias. No additional 

research is proposed at this point by NIOSH. 

SC&A Evaluation: SC&A accepts the NIOSH 2016a explanation because the use of the 

Helgeson mobile whole body counter at Pantex was of concern with respect to measuring 

plutonium and americium, but likely overestimated doses from uranium. As such, the use of 

Helgeson data is claimant favorable as applied to NUMEC, where it was used to estimate 

uranium body burdens. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.12 FINDING 12: REVISION OF TABLE 6-2 AND ASSOCIATED TEXT IN 

SECTION 6.3.2 OF THE SITE PROFILE

Finding 12: Table 6-2 and the associated text in Section 6.3.2 of the site profile 

should be reviewed and modified as needed to correct any oversights, 

inconsistencies, or errors. 

Proposed NIOSH resolution: The indium foil criticality dosimeters were not 

included in Table 6-2 because they were not used for routine workplace 

exposures. They were included only to determine dose from a criticality in the 

event that a criticality occurred. No criticality incidents were reported at the 

NUMEC facilities. The text was reviewed as suggested and the information in 

the text is consistent with information provided in Tables 6-2 and 6-3. 

August 2015 WG discussion: SC&A insists there is still an error in the table. 

How are data from multi component badges handled? SC&A to elaborate 

further. 

SC&A response after WG discussion: More information is needed regarding 

how the data from the neutron detection devices will be used to reconstruct 

neutron doses. For example, Table 6-2 describes the method for deriving the 

fast neutron dose that involved subtraction of the thermal dose, but it is not 

clear if this was determined from a cadmium-filtered film badge or not. The 

Landauer “Z1” dosimeter contains a neutron sensitive CR-39 component, but 

this is not mentioned in the table or text. Both the text and table are unclear 

and inconsistent regarding the use of CR-39. Table 6.2 indicates that Z1 

badges are used (for beta/gamma), but nowhere does it say that they contain a 

CR-39 neutron component. The text in section 6.3.2 mentions the use of NTA 

[nuclear track emulsion, Type A], then states that, after 1968, neutron 

monitoring was performed with TLDs [thermoluminescent dosimeters]. 

Table 6-2 does say that other types of neutron dosimeters (not mentioned in 

the body of the text) did contain CR-39 components. NIOSH needs to revisit 

this section and ensure that both the table and text agree with each other and 

with the dosimetry practices of the period 
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NIOSH update June 2016: The available guidance is suitable for assigning 

neutron doses for partial dose reconstruction of external neutron dose. 

NIOSH will clarify any outstanding questions from SC&A during the 

upcoming work group discussion. 

SC&A Evaluation: It is important to document as accurately as possible the types and 

capabilities of dosimetry utilized throughout the site history. Two issues remain unresolved, 

especially about neutron dosimetry. The first is to accurately reflect historical practice in the site 

profile prose descriptions and summary tables. The second is to address the practices that were 

followed by the dosimetry service providers and NUMEC for determining the recorded dose 

where multiple neutron dosimetry components were utilized for a given individual. As the 

technology changed, so did energy response thresholds, minimum and maximum reportable 

doses, and overlap between devices detecting differing portions of the neutron spectrum. This 

needs to be laid on top of a complex site with changing processes, personnel, and facilities.

The site utilized the entire wide-ranging arsenal of Landauer neutron dosimetry capabilities, 

which is an indication that there were complex neutron measurement issues and that radiation 

safety personnel were attempting to do the best possible job. Technologies provided by Landauer 

included cadmium filtered film for thermal neutrons, TLD 600 and 700, CR39 track etch, NTA 

film, and electrochemically etched polycarbonate. In addition, neutron-sensitive TLD 100 or 

insensitive 700 chips were incorporated into photon dosimeters depending on whether a neutron 

field was anticipated. In some cases, more than one of these neutron dosimeters were worn at the 

same time in the same holder and, thus, it is important to understand how records were 

maintained to ensure that doses were not duplicated or missed. It is also important to understand 

the rules that both Landauer and NUMEC used in determining which dosimeters were utilized 

and how multiple neutron components assessing differing parts of the neutron spectrum were 

recorded. Likewise, rules for missing or inconsistent component results should be identified. 

In cases where Landauer provided a single dosimeter to measure whole body dose, then the 

recordkeeping was probably standard. However, if multi-component neutron whole body 

dosimeters were worn, then care needs to be exercised in examining the record to ensure that the 

appropriate results were documented. In some cases in Table 6-2, it is not clear how this issue is 

handled; indeed, it is not mentioned. For example, a Landauer “Z” dosimeter is designed to 

measure photon, beta, and neutron radiations. The table omits mention of the neutron capability 

entirely. The Landauer “K” dosimeter is similar to the Z, but without neutron capability. 

However, the K dosimeter was intended for use where neutron fields were not anticipated, so 

standard TLD 100 chips were used. As the Z badge was intended for neutron fields and 

contained a CR39 component, TLD 700 was used for gamma monitoring to prevent neutron 

sensitivity. Obviously, it is important to document exactly the dosimeters used or dose 

reconstruction and audit personnel may be lead to false conclusions. Section 6.3.2 of Revision 03 

to the site profile is incorrect, or inconsistent with Table 6-2, as it fails to mention CR39 and 

electrochemically etched polycarbonate dosimetry:

6.3.2 Neutron Dosimeters 

Workers were monitored for neutron exposures with nuclear track emulsion, 

Type A (NTA) film from commercial vendors until about 1968 and with TLDs 
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thereafter. In addition, criticality dosimetry monitoring was done with an array of 

area critical assemblies that fed into a central system. This system existed from at 

least 1963; in September 1963, each visitor and employee was issued an indium 

foil criticality dosimeter as part of each security badge (NUMEC 1963). 

No mention is made of CR39 or electrochemically etched dosimeters. If dose reconstruction 

personnel read this paragraph, they will be confused as to actual practices on the sites. One is left 

to draw the conclusion that the (admittedly very wide) selection of dosimetry that was in use is 

not fully appreciated. Compare this with Table 6-2, excerpted here in Table 1: 

Table 1: Neutron Dosimeters – WB 

Period Monitoring Technique Dosimeter Description

1957–05/1968 NTA film badge Film badges using NTA films: Fast neutrons undergoing elastic 

collision with content of emulsion or cellulose acetate base 

material produce recoil protons, which are recorded as 

photographic tracks in emulsion. Track density is a linear function 

of dose. Developed image exhibits tracks caused by neutrons, 

which can be viewed using appropriate imaging method (i.e., oil 

immersion) and 1,000-power microscope or projection capability. 

06/1968–1995 Landauer Neutrak 

Extended Range 

dosimeter (types 18, I1, 

or RI)

Combined TLD albedo neutron monitor with track recoil device 

[CR-39 (allyl diglycol carbonate)] that responds to neutron 

radiation through proton recoil events. The dosimeter is 

responsive to a neutron energy range of about 0.0001 to 10 MeV. 

Dosimeter response to thermal neutron radiation was subtracted to 

yield fast neutron dose. The Neutrak ER has an albedo element 

with above-described elements. Qualitative relationship was 

derived to determine ratios of neutrons of various energies. The RI 

badge was capable of monitoring beta, X-ray, gamma, and 

neutrons.

07/1991–

12/1991

Teledyne Isotopes TLD 

badge

Combined gamma, beta, and neutron TLD (BWXT 1991). Details 

of the dosimeter are not available, other than detection limits. 

Note: excerpted from Table 6-2 of NIOSH 2016c. 

Given the above, it is recommended that the TBD be modified to provide a more complete 

picture of the neutron dosimetry program and to address how to handle a wide range of potential 

anomalies and exceptions.

SC&A Recommendation: SC&A recommends that this finding remain open until a more 

complete picture of the neutron dosimetry program is provided and the TBD addresses how to 

handle a wide range of potential anomalies and exceptions. 

2.13 FINDING 13: BETTER DEVELOP DESCRIPTION OF THE SOURCES AND 

CIRCUMSTANCES RESPONSIBLE FOR EXTERNAL EXPOSURES 

Finding 13: Given our understanding that it is NIOSH’s position that external 

exposures at the Parks Township Site can be reconstructed with sufficient 

accuracy, it appears that the description of the sources and circumstances 

responsible for external exposures need to be better developed, if possible. 
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Proposed NIOSH resolution: External dose can be evaluated only when 

dosimetry records exist. There was one dosimetry department for all NUMEC 

facilities that provided dosimetry for both the Parks Township and Apollo 

facilities. The limitations stated for the Apollo facilities also apply to the 

Parks Township facilities (see response to Finding 16). 

August 2015 WG discussion: NIOSH to issue response on whether an external 

coworker model is feasible. 

NIOSH update June 2016: It is not expected that the available data can be 

used to develop a coworker model that would adhere to the guidelines 

expected from such a model. Please see detailed analysis below. 

SC&A Evaluation: SC&A reviewed the attachment to NIOSH 2016a with respect to being able 

to construct an external dosimetry coworker model and notes that it provides a detailed 

description of the quality and completeness of both the internal bioassay and external dosimetry 

data. SC&A agrees with NIOSH’s conclusion that the available external dosimetry data cannot 

support the development of an external dosimetry coworker model. Section 6.0 of Revision 03 to

the TBD states that external doses will be assigned to workers to the extent that external 

dosimetry was provided for a given worker.

SC&A Recommendation: SC&A recommends closing this issue. 

2.14 FINDING 14: NUCLEAR TRACK EMULSION TYPE A (NTA) FILM

ADJUSTMENT FACTORS FOR NEUTRON EXPOSURES

Finding 14: The site profile should provide justification for why adjustment 

factors are not required for neutron exposures estimated using nuclear track 

emulsion type A (NTA) film, considering that it appears that the neutron energy 

spectrum likely extended to well below 1 mega-electronvolt (MeV). For example, 

Table 6-8 of the site profile indicates that the energy range of neutron exposures 

extended from 0.1 to 2 MeV. 

Proposed NIOSH resolution: A n:p [neutron-to-photon dose ratio; also “n/p”] 

approach was developed from available data. 

August 2015 WG discussion: SC&A states that NIOSH response is thorough, 

but is not sure it makes technical sense. SC&A is to provide a response after 

additional review. 

SC&A response after WG discussion: While the n/p methodology that NIOSH 

is proposing for NUMEC has been used at other AWE and DOE sites, the 

dose data used to derive the n/p values suggested by NIOSH for NUMEC are 

not comprehensive or robust. NIOSH may want to consider adopting a 

bounding n/p ratio that ensures that the neutron doses are not understated for 

circumstances where the available n/p data are limited. 
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NIOSH update June 2016: NIOSH has reviewed the suggested approach, and 

agrees that the data are limited; however, NIOSH is not aware of any 

additional data that may exist and that would allow for refining the currently 

proposed approach in any way. The proposed approach was included in draft 

ORAUT-TKBS-0041 Rev. 03-B.

SC&A Evaluation: NIOSH 2015b states that NUMEC used NTA film to monitor for neutron 

exposure until approximately June 1968, when TLDs were introduced at the site. NIOSH 

explains that a study was performed in 1975 using TLDs to determine the photon and neutron 

doses while working in the FFTF fuel fabrication area of the Parks Township plutonium facility. 

Additionally, dosimeters were placed at various fixed locations. The NIOSH study resulted in the 

following estimate of neutron-to-photon ratio (Author unknown 1977; also, BWTX 1975, 

BWTX 1981, and BWTX 1982), shown in Table 2.

Table 2. Neutron-to-Photon Ratios on Personnel Dosimeters

Geometric Mean 0.34

GSD 1.71

Source: Table copied from the Finding 14 section of NIOSH 2015b. 

This ratio is supported by the Health and Safety ALARA Reports BWTX 1981 and BWTX 1982, 

which indicate that neutron-to-photon ratios (n/p) varied from 0.23 to 0.42 from 1979 to 1981, 

with an average of 0.33. The geometric mean (GM) n/p value of 1.00 for glovebox workers was 

obtained from BWTX 1975 (Table 6, PDF page 27) from the 42 reported 1975 photon-toneutron dose ratio (p/n) values (converted to n/p values). 

Table 3. Neutron-to-Photon Ratio Glovebox Dosimeters 

Geometric Mean 1.00

GSD 1.49

Source: Table copied from the Finding 14 section of NIOSH 2015b.

In addition, NIOSH stated that information was obtained from a September 1968 event in which 

a worker involved with manufacturing neutron sources had a neutron-to-photon ratio of 2.33 

(determined using estimated neutron dose values) (Caldwell 1968). A review of neutron-tophoton ratios from neutron sources was evaluated and a broad range of ratios were found. 

Without additional information to indicate otherwise, NIOSH adopted a ratio of 2.33. 

On this basis, NIOSH proposed that NUMEC workers, prior to 1969 and with neutron dosimetry, 

receive a neutron dose assigned using the ratio most appropriate for their work and job location 

(i.e., use a ratio of 0.34 for typical workers, a ratio of 1.00 for glovebox workers, and a ratio of 

2.33 for workers involved with manufacturing sources). This dose would be assigned using a 

lognormal distribution with the GSD provided, except for the manufacturing source ratio, which 

would be applied as a constant. If the worker’s recorded neutron dose is higher than the neutron 

dose calculated from applying the ratio to the photon dose, the recorded neutron dose should be 

assigned. NIOSH indicated that this is the approach it will use to reconstruct neutron doses for 

workers prior to 1968, the time period when NTA film was used, and the revised TBD will 

incorporate this new information. 
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During the WG meeting in August 1015, SC&A expressed some concern regarding the sources 

and robustness of the data used to derive the n/p ratios. Given this background information, 

SC&A reviewed Revision 03 to the TBD to determine the degree to which NIOSH fulfilled its 

commitments.

SC&A’s review of Section 6.5.2.4 reveals that NIOSH did, in fact, incorporate the new material 

discussed in theory in the white paper and during WG meetings. In addition, SC&A reviewed the 

source documents upon which NIOSH based its recommended n/p ratios as used in Revision 03 

to the TBD and found the following: 

1. n/p = 0.34: The n/p GM value of 0.34 was derived from averaging the gamma and 

neutron TLD data in the three columns in Table 1, page 26, of BWTX 1975. These data 

consisted of recorded neutron and gamma doses for 17 operators and associated workers 

for three different months in 1975. 

2. Supporting recommendation of n/p = 0.34: To support the recommended n/p value of 

0.34, NIOSH showed that other data provided a similar n/p value of 0.33, derived from 

the n/p values of 0.23 to 0.42, with an average of 0.33. These data were obtained from 

BWTX 1981, which lists the 1979 average n/p value as 0.42 and the 1980 average n/p 

value as 0.35, and BWTX 1982, which lists the 1980 average n/p value as 0.35 and the 

1981 average n/p value as 0.23. The overall average n/p values would be (0.42 + 0.35 + 

0.23)/3 = 0.33. 

3. n/p = 1.00: The GM n/p value of 1.00 for glovebox workers was obtained from BWTX 

1975, Table 6, PDF page 27, from the 42 reported 1975 p/n values (converted to n/p 

values). SC&A calculated an n/p GM value of 0.94 and an average n/p value of 1.08, 

which closely matches NIOSH’s recommended value of 1.00. 

4. n/p = 2.33: The source worker’s n/p value of 2.33 was obtained from one worker 

involved with manufacturing neutron sources in 1968 (Caldwell 1968). SC&A found that 

this n/p value was derived from the first quarter exposure results for 1968 for a worker 

who has a 1.309 rem photon TLD dosimeter reading and a total neutron dose of 

3.054 rem calculated from a combination of time-in logs and area neutron surveys 

(n/p = 3.054/1.309 = 2.33). 

SC&A’s research into the source documents used by NIOSH to derive the n/p ratios reveals that 

they appear to be scientifically sound. However, since all the values were obtained from 

snapshots in time, we are concerned whether they are representative of the complex operations 

and long period of time that workers at NUMEC may have experienced neutron exposures. For 

example, both NUMEC locations performed a variety of handling and fabrication operations 

with uranium and plutonium, from reactor fuels to (alpha, n) neutron sources, with some work 

conducted in gloveboxes. These varied operations present a challenge in reconstructing neutron 

doses, especially compared to an AWE facility that only performed routine fuel processing.

While the n/p methodology that NIOSH is proposing for NUMEC is plausible, the dose data 

used to derive the n/p values suggested by NIOSH for NUMEC are not as robust as would be 

desired, for the following reasons: 

1. The n/p = 0.34 value to be applied to typical workers was derived from one 3-month 

study conducted in 1975 for 17 workers at one location (Table 1 of BWXT 1975). This 
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would only provide a snapshot in time of the n/p value and would not necessarily be 

representative of all the typical workers’ locations and time periods where NTA film was 

used (1950s and 1960s) at the two NUMEC locations. 

2. The n/p = 1.00 value to be applied to glovebox workers was derived from one 3-month 

study conducted in 1975 for personnel working near 20 gloveboxes (Table 6 of BWXT 

1975). This would only provide a snapshot in time of the n/p value and would not 

necessarily be representative of all the typical gloveboxes and time periods where NTA 

film was used (1950s and 1960s) at the two NUMEC locations. 

3. The n/p = 2.33 value to be applied to workers involved in manufacturing neutron 

sources was derived from one 3-month gamma TLD dosimeter reading and area neutron 

surveys using neutron survey instruments and a time log for one worker in 1968 

(Caldwell 1968). Details of the processing situation (gloveboxes, shielding, etc.) were not 

provided. This would only provide a snapshot in time of the n/p value and would not 

necessarily be representative of all the neutron source manufacturing exposures and time 

periods where NTA film was used (1950s and 1960s) at NUMEC. 

In order to evaluate the degree to which the n/p ratios are reasonable as applied to NUMEC 

operations, SC&A reviewed n/p ratios from two other DOE sites that performed similar 

operations processing uranium and plutonium. The following is a summary of our investigations: 

1. Rocky Flats Plant – ORAUT-TKBS-0011-6, Revision 03, Technical Basis Document 

for the Rocky Flats Plant (NIOSH 2010a), Table 6-21, lists n/p values for the period 

1970–1976 that range from 0.67 to 1.61, with an average n/p value of 1.05. Additionally, 

Table 6-22 lists n/p values for the period 1977–2000 that range from 0.26 to 0.61, with an 

overall n/p value of 0.42. 

2. LANL – ORAUT-TKBS-0010-6, Revision 03, Technical Basis Document for the Los 

Alamos National Laboratory – Occupational External Dose (NIOSH 2013), lists the 

following n/p information: 

• Table 6-16 lists n/p values from a 1972 study that range from 1.33 to 3.98 for the 

depleted plutonium site. 

• Table 6-18 lists n/p values from a 1968 study that range from 0.67 to 2.8 as a 

function of distance from a critical assembly.

• Table 6-22 lists overall recommended LANL n/p values for various 

facilities/operations at LANL that range from 0.7 to 3.9. 

• Table B-1 lists an overall n/p value of 1.04 for 1950–2008 from LANL recorded 

dose data.

• Table B-2 lists overall n/p values ranging from 0.53 to 2.66 for the time period 

1957–1968 from LANL recorded doses. 

With the additional reference information provided in the revised NUMEC TBD, SC&A was 

able to derive n/p values as recommended by NIOSH for NUMEC and obtain more details 

concerning their sources and applicability. While the NUMEC n/p data set is not as robust as 

desired because it is limited to providing only a snapshot of the situation in time and location, the 
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derived NUMEC n/p values are not out of line with other DOE facilities that handled uranium 

and plutonium. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.15 FINDING 15: NON-PENETRATING DOSE ASSIGNMENTS USING FILM 

BADGE DOSIMETERS

Finding 15: The markedly different photon energies associated with the 

operations at NUMEC would indicate the possible need for adjustment factors for 

the results of film badge dosimeters, which are not provided in the site profile. 

Proposed NIOSH resolution: Film badge dosimeters, while over responding 

to radiation recorded in the open window, may under respond to low energy 

photons (16 keV and 59 keV photons are a particular concern) (Wilson et al. 

1990). Although the films and filters at NUMEC were different than the 

dosimeters discussed in the reference, a reasonable comparison between the 

film dosimeters is expected (AEC 1955). The site acknowledged this deficiency 

in a 1966 report (Caldwell and Judd 1966) and made corrections to the 

dosimetry to account for this under response. Prior to the report being issued 

(i.e. prior to 1966), to account for under response of film dosimetry to low 

energy photons, the result in the open window should be assigned as <30 keV 

photons for workers at plutonium facilities while the deep dose response is 

assigned in accordance with the worker location. 

August 2015 WG discussion: Auger electrons are very low energy and would 

not penetrate skin; therefore, part of NIOSH response does not address this 

issue. Also, do we assign these low energy betas for all alpha emitters at other 

sites? NIOSH wants to check this answer and get back, but WG decided that 

this is an SC&A action item for now – SC&A is to provide additional 

information on this finding. 

SC&A response after WG: NIOSH concurred that the site profile needs to be 

revised to account for possible over- and/or under-responses of dosimeters 

under some circumstances. The decision to modify the site profile approach 

for situations where low-energy photons and betas are present seems 

appropriate by assuming <30 keV photons for plutonium workers for open 

window results. However, NIOSH should endeavor to gather more 

information on the differing film badges and the protective coverings to 

enable a better assessment of situations where low-energy photons or betas 

were under-reported or missed entirely. 

NIOSH update June 2016: The initially proposed guidance was added to the 

draft ORAUT-TKBS-0041 Rev. 03-B. Additional data capture is considered to 

be unproductive, but additional review in available NUMEC files continues. 
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SC&A Evaluation: SC&A reviewed Section 6.5.2.1 and Table 6-5 of Revision 03 to the TBD, 

and NIOSH does, in fact, recommend assigning non-penetrating dose as <30 keV photons for 

plutonium workers and as >15 keV electrons for other workers as per ORAUT-OTIB-0017, 

Revision 01, Interpretation of Dosimetry Data for Assignment of Shallow Dose (NIOSH 2005). 

SC&A’s additional observation that it is unclear what the authors intended by describing all beta 

energies for Am-241 as being >15 keV, as Am-241 and U-233 are alpha emitters, is also 

resolved. 

SC&A Recommendation: SC&A recommends closing this issue 

2.16 FINDING 16: NUMEC COWORKER MODEL

Finding 16: NIOSH should consider developing a universal coworker model 

based on NUMEC claimant records, or specify a more consistent basis for 

assigning external doses beyond the medical x-rays associated with the site. 

Proposed NIOSH resolution: A coworker model for non-presumptive claims is 

not expected to be feasible. 

August 2015 WG discussion: NIOSH is tasked to do a formal review to see if a 

coworker model is feasible. 

NIOSH update June 2016: It is not expected that the available data can be 

used to develop a coworker model that would adhere to the guidelines 

expected from such a model. Please see detailed analysis below. 

Following discussion of the 21 findings in NIOSH 2016a, NIOSH provides its rationale for 

determining that it is not feasible to develop a coworker model for workers with nonpresumptive cancers at NUMEC. Internal exposure to uranium at Apollo and external exposures 

at Parks Township are of particular interest. The white paper investigated this matter by 

researching the extent to which uranium bioassay data were available for workers at Apollo and 

external dosimetry data that were available for workers at Parks Township. NIOSH found that 

some non-covered workers had such data available, and those data were used to reconstruct 

partial dose reconstructions for those workers, where feasible. However, NIOSH determined that 

the amount, completeness, and quality of the data did not meet NIOSH’s criteria for developing 

coworker models as set forth in NIOSH 2015c. 

NIOSH 2016a provides an extensive discussion of the types of data that are available as well as 

those that are not available. For example, the white paper explains that the uranium bioassay data 

that are available do not provide information regarding the location of the worker (e.g., Apollo 

versus Parks Township). With respect to Parks Township, NIOSH 2016a compiles and presents 

the external dosimetry data that are available (i.e., 28,000 lines of data). The white paper 

explains that data are missing for some years, some of the data were for workers involved in 

commercial operations, some data lacked job and location indicators, and some workers clearly 

worked in areas where the potential for exposures existed but were not monitored. On the bases 

of these findings, NIOSH 2016a concluded that the data have many deficiencies and could not be 

used to construct an external dosimetry coworker model. 
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This subject was discussed at the WG meeting on July 19, 2016 (beginning on page 25 of the 

transcript). The key point that emerged in this discussion begins on page 118 of the transcript. 

The point is made that although there are considerable external dosimetry data, there are data 

gaps that indicate that some of the high exposures might be missing. The basis for this judgment 

is that many workers were monitored for internal exposures but did not have external dosimetry. 

It is then reasonable to conclude that, without assurance that high-end exposures were captured, 

it might not be possible to develop an external coworker model. In addition, the point was made 

that in many cases it cannot be determined if the data apply to workers at Apollo or Parks 

Township. 

An important policy discussion followed that SC&A believes should be documented here. As 

described in the WG transcript beginning on page 130, although there might be a fairly large 

external dosimetry data set for a given year, if there is some question that the data set does not 

capture the upper bound or that the locations of the exposures are not known, the data set does 

not meet the criteria for building a coworker model. As articulated by Mr. Katz (the Designated 

Federal Officer), it has long been NIOSH’s policy that minimal exposures are not to be assigned 

when developing coworker models. It is essential that the coworker model be used to assign 

realistic or at least plausible bounding exposures. It is clear that the external dosimetry data at 

Parks Township do not meet this test, and, on this basis, it would not be appropriate to try to 

construct an external coworker model. 

SC&A Evaluation: SC&A accepts the above NIOSH response that it is not feasible to construct 

a coworker model for non-presumptive cancers. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.17 FINDING 17: NON-PENETRATING DOSES FROM BETA EMITTERS 

ASSOCIATED WITH RESIDUAL PERIOD SURFACE CONTAMINATION

Finding 17: The site profile should include guidance for deriving non-penetrating 

doses to skin and other organs from beta emitters associated with surface 

contamination during the residual period. 

Proposed NIOSH resolution: Additional information was included in the draft 

of the revised TBD. 

August 2015 WG discussion: Closed. 

SC&A Evaluation: Section 7.1 of Revision 03 to the TBD addresses the reconstruction of doses 

during the residual period. Table 7-1 and its supporting text provides detailed guidance regarding 

the reconstructing of external dose from surface contamination, including skin doses from beta 

contamination. 

SC&A Recommendation: SC&A recommends closing this issue. 
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2.18 FINDING 18: GENERAL AIR SAMPLES TO RECONSTRUCT RADIONUCLIDE 

INTAKE RATES DURING THE RESIDUAL PERIOD

Finding 18: General air (GA) samples, as opposed to breathing zone (BZ) 

samples, should be used as the starting point for reconstruction of radionuclide 

intake rates during the residual period. 

Proposed NIOSH resolution: The air concentration value that is favorable to 

claimants to use for the residual period is represented by the maximum 

median value of 222 dpm/m3. The maximum estimated GSD for all data sets is 

6.95 for the HASL GA data. Therefore, the residual activity will be based on a 

lognormal distribution with a median of 222 dpm/m3 and a GSD of 5.0. This 

represents a slight increase in median air concentration from the previous 

concentration of 210 dpm/m3

 and GSD of 7.91. 

August 2015 WG discussion: Some BZ samples are higher than some of the 

GA samples. Why is that? Possibly there are some process samples that were 

added to the GA sample set. The data need to be reviewed to make sure there 

are no inconsistencies. 

NIOSH update June 2016: The data were reviewed and corrected if necessary 

and the revised approach was included in draft ORAUT-TKBS-0041 

Rev. 03-B. The results from these measurements and the HASL reports 

indicate an air concentration value that is favorable to claimants to use for 

the residual period is represented by the maximum median value of 

329 dpm/m3

. 

SC&A Evaluation: Inspection of Section 7.4.1 of Revision 03 to the TBD reveals that NIOSH 

plans to use the maximum median air concentration of uranium of 328 dpm/m3 as the median 

concentration, with a GSD of 6.95, as the starting point for deriving the uranium intake rate for 

workers during the residual period. This is a reasonable bounding airborne uranium dust loading 

as the starting point for reconstructing internal doses during the residual period because it is 

based on measurements made at several locations at the facility during operations. Hence, it 

represents an upper bound as applied to the residual period. This value is used to derive the 

concentration of uranium deposited on surfaces, which is used to estimate airborne dust loading 

using a resuspension factor of 10-5/m, along with other assumptions described in ORAUT-OTIB0070, Revision 01, Dose Reconstruction during Residual Radioactivity Periods at Atomic 

Weapons Employer Facilities (NIOSH 2012b). The entire protocol is in accord with all 

previously approved procedures for reconstructing doses during the residual period at facilities 

that handled uranium. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.19 FINDING 19: RESUSPENSION FACTOR DURING THE RESIDUAL PERIOD

Finding 19: SC&A recommends that NIOSH use a resuspension factor of about 

1E-5 per meter to derive the airborne dust loading for the beginning of the 
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residual period, or perhaps simply assume that the average general air dust 

loading observed during the residual period is applicable to the beginning of the 

residual period. 

Proposed NISOH resolution: Resuspension factor for Apollo was changed, 

but not for Parks, because of cleanup activities. 

August 2015 WG discussion: Closed, if the guidance is included in the TBD 

revision. 

NIOSH update June 2016: The change was included in draft ORAUT-TKBS0041 Rev. 03-B.

SC&A Evaluation: See Finding 18 and SC&A’s associated response and recommendations. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.20 FINDING 20: RADIONUCLIDES OTHER THAN URANIUM DURING THE 

RESIDUAL PERIOD AT APOLLO

Finding 20: The site profile makes no reference to radionuclides other than 

uranium during the residual period at Apollo 

Proposed NIOSH resolution: Suggestion for adding thorium and progeny to 

approach. 

August 2015 WG discussion: Closed, if the thorium guidance is added to the 

TBD revision. 

NIOSH update June 2016: The change was included in draft ORAUT-TKBS0041 Rev. 03-B.

SC&A Evaluation: Section 5.2.3, “Thorium Exposures,” of Revision 03 to the TBD explains 

that thorium was processed at NUMEC at both Apollo and Parks Township sites beginning in the 

early 1960s, and that some thorium bioassay data are available. However, the TBD explains that 

not all workers who should have been monitored for thorium were in fact on a routine thorium 

monitoring program, nor are there adequate data characterizing the airborne concentrations for 

thorium. Therefore, the TBD concludes that it is not possible to construct a coworker model for 

thorium exposures. However, the TBD explains that, for workers who have thorium bioassay 

data, the internal doses to those workers from the intake of thorium will be reconstructed. 

SC&A Recommendation: SC&A recommends closing this issue. 

2.21 FINDING 21: GUIDANCE ON TREATING AGED PLUTONIUM MIXTURES 

DURING THE RESIDUAL PERIOD AT PARKS TOWNSHIP

Finding 21: There is conflicting guidance on how aged plutonium mixtures 

should be treated during the residual period at Parks Township. 
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Proposed NIOSH resolution: add guidance to TBD 

August 2015 WG discussion: Closed, if guidance was added to the TBD 

revision. 

NIOSH update June 2016: The change was included in draft ORAUT-TKBS0041 Rev. 03-B.

SC&A Evaluation: The Publication Record of Revision 03 to the TBD states that revised 

Section 5.1 and Table 5-3 address this issue. Inspection of the TBD reveals that Table 5-3 

provides detailed guidance on the mix of plutonium isotopes that should be assumed for different 

grades and ages of plutonium. 

SC&A Recommendation: SC&A recommends closing this issue. 
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3.0 REFERENCES

Author unknown 1977. Summary of B&W and NRC Review Team Meeting, January 19, 1977. 

[SRDB Ref. ID 142792] 

Author unknown, ca. 2004. NUMEC Apollo Facility. [SRDB Ref. ID 30958] 

AEC 1955. Intercomparison of Film Badge Interpretations, U.S. Atomic Energy Commission, 

Isotopics, Volume 5, Number 2, pp. 8–33. April 1955. [SRDB Ref. ID 3640]

Brake 1989. R. Brake, Los Alamos National Laboratory, letter to G. Runkle, U.S. Department of 

Energy. November 9, 1989. [SRDB Ref. ID 11091] 

BWXT 1975. Radiation Control Technology Progress Report, Babcock & Wilcox Technologies, 

Apollo, Pennsylvania. December 1975. [SRDB Ref. ID 142792, PDF p. 19] 

BWXT 1981. 1980 Health and Safety ALARA Report, Babcock & Wilcox Technologies, Apollo, 

Pennsylvania. April 22, 1981. [SRDB Ref. ID 142785]

BWXT 1982. 1981 Health and Safety ALARA Report, Babcock & Wilcox Technologies, Apollo, 

Pennsylvania. March 23, 1982. [SRDB Ref. ID 142790] 

Caldwell 1968. Letter from R.D. Caldwell, Nuclear Materials and Equipment Corporation, to 

L.D. Low, U.S. Atomic Energy Commission, Division of Compliance, Subject: Excessive 

Exposures SNM-414. September 23, 1968. [SRDB Ref. ID 142650, PDF pp. 91–92] 

Caldwell 1969. Letter from R.D. Caldwell, Nuclear Materials and Equipment Corporation, to 

L.D. Low, U.S. Atomic Energy Commission, Subject: Excessive Exposures, SNM-414 and 

SNM-145, October, 1969. November 4, 1969. [SRDB Ref. ID 19970] 

Caldwell and Judd 1966. R.D. Caldwell and W.C. Judd, Gamma Spectrum Measurements and 

the Interpretation of Absorbed Dose during Plutonium Fuel Fabrication, Nuclear Materials and 

Equipment Corporation, Apollo, Pennsylvania, paper presented at the 1966 Meeting of the 

Health Physics Society, Houston, Texas. 1966. [SRDB Ref. ID 22776]

Crosby 1967. E. Crosby, Individual Dosimetry Data Plutonium Facility 1967. [SRDB Ref. 

ID 19734] 

Davis and Strom 2008. A.J. Davis and D. J. Strom, “Uncertainty and Variability in Historical 

Time-Weighted Average Exposure Data,” Health Physics, Vol. 94, No. 2 (2008), pp.145–160. 

[SRDB Ref. ID 41616] 

ICRP 1995. International Commission on Radiological Protection, Age-Dependent Doses to 

Members of the Public from Intake of Radionuclides – Part 4 Inhalation Dose Coefficients, 

Publication 71, Ann. ICRP 25 (3-4). 1995. [SRDB Ref. ID 22729] 
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NIOSH 2004. Technical Basis Document for the Fernald Environmental Management Project 

(FEMP) – Occupational Internal Dose, ORAUT-TKBS-0015-5, Revision 00, National Institute 

for Occupational Safety and Health, Cincinnati, Ohio. May 28, 2004. 

NIOSH 2005. Interpretation of Dosimetry Data for Assignment of Shallow Dose, ORAUTOTIB-0017, Revision 01, National Institute for Occupational Safety and Health, Cincinnati, 

Ohio. October 11, 2005. 

NIOSH 2007a. Fission and Activation Product Assignment for Internal Dose-Related Gross Beta 

and Gross Gamma Analyses, ORAUT-OTIB-0054, Revision 00 PC-1, National Institute for 

Occupational Safety and Health, Cincinnati, Ohio. November 19, 2007. 

NIOSH 2007b. Pantex Plant – Occupational Internal Dose, ORAUT-TKBS-0013-5, 

Revision 01, National Institute for Occupational Safety and Health, Cincinnati, Ohio. June 22, 

2007. 

NIOSH 2008. Site Profile for Nuclear Materials and Equipment Corporation, Apollo and Parks 

Township, Pennsylvania, ORAUT-TKBS-0041, Revision 0, National Institute for Occupational 

Safety and Health, Cincinnati, Ohio. September 26, 2008. 

NIOSH 2010a. Technical Basis Document for the Rocky Flats Plant – Occupational External 

Dose, ORAUT-TKBS-0011-6, Revision 02 PC-1, National Institute for Occupational Safety and 

Health, Cincinnati, Ohio. October 20, 2010. 

NIOSH 2010b. Technical Basis Document for the Hanford Site – Occupational External Dose, 

ORAUT-TKBS-0006-6, Revision 04, National Institute for Occupational Safety and Health, 

Cincinnati, Ohio. January 7, 2010. 

NIOSH 2010c. Estimating Doses for Plutonium Strongly Retained in the Lung, ORAUT-OTIB0049, Revision 01 PC-2, National Institute for Occupational Safety and Health, Cincinnati, Ohio. 

November 29, 2010. 

NIOSH 2012a. Site Profile for Nuclear Materials and Equipment Corporation, Apollo and Parks 

Township, Pennsylvania, ORAUT-TKBS-0041, Revision 02, National Institute for Occupational 

Safety and Health, Cincinnati, Ohio. November 26, 2012. 

NIOSH 2012b. Dose Reconstruction during Residual Radioactivity Periods at Atomic Weapons 

Employer Facilities, ORAUT-OTIB-0070, Revision 01, National Institute for Occupational 

Safety and Health, Cincinnati, Ohio. March 5, 2012. 

NIOSH 2013. Technical Basis Document for the Los Alamos National Laboratory – 

Occupational External Dose, ORAUT-TKBS-0010-6, Revision 03, National Institute for 

Occupational Safety and Health, Cincinnati, Ohio. March 21, 2013. 

NIOSH 2015a. Fission and Activation Product Assignment for Internal Dose-Related Gross Beta 

and Gross Gamma Analyses, ORAUT-OTIB-0054, Revision 04, National Institute for 

Occupational Safety and Health, Cincinnati, Ohio. August 27, 2015. 
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Effective Date:

2/6/2017

Revision No.

0 (Draft)

Document No./Description:

SCA-TR-2017-SP001

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Privacy Act 5 U.S.C. § 552a and has been cleared for distribution.

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